Reprocessing of spent nuclear fuel can be used to recover plutonium, uranium, and other materials. The current standard method for fuel reprocessing is the PUREX method, which is a liquid-liquid extraction method that can extract both uranium and plutonium independently of each other as well as from other fission products. The liquid-liquid extraction process generates high volumes of liquid waste that carries many of the fission products and transuranic elements generated in the core as well as actinides and a plurality of different salts. The raffinate from the PUREX method is treated to separate a portion of the waste, which is then precipitated and vitrified. The resulting glass is retained in long-term storage containers. The remaining large volume of liquid waste is further treated to remove actinides and fission products to the greatest extent possible.
The liquid waste treatment processes are generally designed to remove cationic species like cesium (e.g., cesium-137) and strontium (e.g., strontium-90). Unfortunately, certain fission products that form anionic species, such as technetium in the form of pertechnetate, are not removed from the liquid waste by these treatment processes. As such, these anionic fission products often remain in the post-treatment salt solution. This salt solution exhibits low activity levels and must be properly treated and stored to prevent release of remaining contaminants. In addition, the waste acceptance criteria for this solution changes with time, and it is expected that future requirements will call for the more complete removal of contaminants such as technetium from the large volume salt solution.
All of the isotopes of technetium are radioactive, and many of them, particularly those produced as a result of fission reactions, have long half-lives, which raise major concerns for long-term storage. Current treatment of the remaining salt solution includes mixing it with a blend of cementitious materials to form a grout mixture. Upon curing of the grout mixture through hydration reactions, a hardened monolithic cementitious waste form known as saltstone is formed. Saltstone is designed to be stored at long-term storage sites. This particular treatment process is carried out at the Savannah River Site nuclear reservation in South Carolina, USA.
While this treatment method has proven highly effective for some contaminants, saltstone is unlikely to hold technetium successfully for long term storage as Tc-99 has a half-life of 211,000 years. Over time, the technetium in the saltstone can be oxidized to a more soluble state. If the waste then comes into contact with water, the soluble components can leach into the environment, particularly those soluble components such as oxidized technetium compounds that do not adsorb well onto the surfaces of minerals, making them more likely to escape into the environment.
What are needed in the art are less expensive, straight-forward separation methods that can be utilized to recover technetium from highly alkaline solutions.